NRC Generic Letter 1981-24: Difference between revisions
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| issue date = 06/15/1981 | | issue date = 06/15/1981 | ||
| title = NRC Generic Letter 1981-024: Multi-Plant Issue B-56, Control Rods Fail to Fully Insert | | title = NRC Generic Letter 1981-024: Multi-Plant Issue B-56, Control Rods Fail to Fully Insert | ||
| author name = Eisenhut D | | author name = Eisenhut D | ||
| author affiliation = NRC/NRR | | author affiliation = NRC/NRR | ||
| addressee name = | | addressee name = | ||
Revision as of 04:41, 14 July 2019
| ML031210452 | |
| Person / Time | |
|---|---|
| Issue date: | 06/15/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| GL-81-024, NUDOCS 8107310298 | |
| Download: ML031210452 (2) | |
tA REG(,, k UNITED STATES-NUCLEAR REGULATORY
COMMISSION
< £i WASHINGTON, D. C. 20555 June 15, 1981 LICENSEES
OF ALL GE BWR FACILITIES (EXCEPT HUMBOLDT BAY)SUBJECT: MULTI-PLANT
ISSUE B-56, CONTROL RODS FAIL TO FULLY INSERT (Generic Letter No. 81-24)Gentlemen:
By letters dated August 29, 1978 and January 10, 1980, you were requested to provide information on your experience with one or more control rods stopping short of the fully inserted position on scram and then settling back to notch position "02" or six inches short of full insertion.
We have evaluated your responses, along with those provided by other BWR licensees, and have concluded that the problem of control rods inserting only to notch position "102" on scram is attributable to leakage past worn I stop and drive piston seals. We have determined that this problem is a maintenance problem, and is readily correctable by control rod drive over-haul at a subsequent outage. We have further concluded that Technical Specification requirements regarding control rod operability and shutdown margin provide adequate assurance of the capability to place and maintain the plant in a safe shutdown condition.
Consequently, the failure to fully insert events, as described herein, do not represent a significant safety issue.The enclosed safety evaluation provides the basis for this determination.
Sincerely,/ Darrell G. Eisenhut, rector Division of Licensing Office of Nuclear Reactor Regulation Enclosure:
Safety Evaluation cc w/encl: Service List ,/: i10310295
610615 PDR ADOCK 05000010 P PDR
yw BOILING WATER REACTOR LICENSEES Docket No. 50-293 Pilgrim Unit 1 Docket No. 50-245 Millstone Unit 1 Docket No. 50-263 Monticello Docket No. 50-325 Brunswick Unit 1/Docket No. 50-324 Brunswick Unit 2-.9 04 9 Docket No. 50-10 .Dresden 1 Docket No. 50-277 Peach Bottom Unit 2 Docket No. 50-237 Dresden 2 Docket No. 50-278 Peach Bottom Unit 3 Docket No. 50-249 Dresden 3 Docket No. 50-333 FitzPatrick Docket No. 50-254 Quad-Cities Unit 1 Docket No. 50-259 Browns Ferry Unit 1 Docket No. 50-265 Quad-Cities Unit 2 Docket No. 50-260 Browns Ferry Unit 2 Docket No. 50-155 Big Rock Point Docket No. 50-296 Browns Ferry Unit 3 Docket No. 50-409 Lacrosse Swcj.Docket No. 50-271 Vermont Yankee Docket No. 50-321 Edwin I. Hatch Unit 1 I Docket No. 50-366 Edwin I. Hatch Unit 2 Docket No. 50-331 Duane Arnold Docket No. 50-219 Oyster Creek Docket No. 50-220 Nine Mile Point Unit 1 Docket No. 50-298 Cooper Station